Computational investigations of the dynamical characteristics of the core of a periodic pulse reactor in a system with cascade neutron multiplication

The core dynamics of a fast reactor in a cascade reactor system operating in a periodic-pulse regime are examined. A model of a BN-600 fuel element is used as a computational model. Computational studies of the neutron kinetics processes in a fast rector-subcritical assembly system and the thermal dynamics of a fuel element in the core of a periodic-pulse reactor are performed. Estimates are made of the service life of a fuel element operating in a regime with repeating pulses and a number of heat loads that is admissible from the standpoint of the fatigue strength of the element.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 260–269, October, 2004.     

Prospects for the BN-1800 Sodium-Cooled Fast Reactor Satisfying 21st Century Nuclear Power Requirements

The BN-1800 power-generating unit is designed to meet the requirements of the strategy for developing atomic energy in Russia in the first half of the 21st century. The development time is the next 15 years and construction could start after 2020. The design is innovative and includes the development of key new technical solutions as compared with the BN-800 reactor which is now under construction.The new technical solutions are based on the substantial positive experience in operating fast reactors in Russia (~125 reactor·years), specifically the BN-600 reactor. The innovations make it possible not only to solve strategic problems, such as increasing safety, improving ecology (including by burning actinides), and nonproliferation but also to make large improvements in economic performance.     

BN-800 design validation and construction status

Information is presented on the BN-800 design, the second design following BN-600, power-generating unit with a fast reactor. The main stages of the development of the design begun in the 1980s, modified in the 1990s after the Chernobyl accident, and accepted for construction within the government program starting in 2000 are presented. The fundamental differences of BN-800 from BN-600 are characterized, and current R&D work is briefly described. Information is presented on the construction of BN-800 at the Beloyarskaya nuclear power plant, where the BN-600 has been operating since 1980.     

不对称工况对池式钠冷快堆堆芯入口温度的影响

对于池式钠冷快堆,堆芯入口温度是重要的热工参数之一,电厂设计过程中堆芯入口温度的确定受多种因素制约,其中包括不同电厂工况的影响。不对称工况是一种典型的电厂工况,本文以600 MW两环路设计的池式钠冷快堆为研究对象,采用钠冷快堆系统分析程序分析不对称工况对堆芯入口温度的影响。研究结果表明,在所分析的不对称工况下,冷池温度会出现明显的不对称现象,且其中1个环路的冷池温度明显上升。通过分析可知,作为电厂的重要热工参数,在不对称工况下,堆芯入口温度变化的影响主要体现在对冷池内设备的影响上,对电厂整体功能和性能有所影响但不构成该工况下影响电厂功能和性能的关键因素。     

高性能计算分析技术在大型钠冷快堆屏蔽设计中的应用

钠冷快堆堆容器是一体化的池式结构,由众多堆内构件组成且结构复杂,堆芯到生物屏蔽外中子输运过程中各向异性明显且深穿透问题严重,大尺度范围下三维S_N方法计算是制约快堆屏蔽设计的瓶颈。通过将三维SN程序与高性能计算技术相结合,采用并行计算方法可解决快堆堆本体内各向异性的三维深穿透屏蔽问题。本文以中国示范快堆（CFR600）堆本体为研究对象,采用JSNT-CFR程序详细计算了堆本体内的中子注量率、光子注量率、剂量率,并将计算结果与已有的二维程序设计结果进行比较。结果表明,将传统屏蔽计算方法与高性能计算相结合,能满足CFR600堆本体屏蔽计算精度要求,获得更为全面的三维展示效果,在计算模型复杂、粒子穿透深度等复杂问题的屏蔽计算上具有较明显的优势,为大型钠冷快堆屏蔽设计提供有力支撑。     

Application of High Performance Computing Analysis Technology in Shielding Design of Large Sodium-cooled Fast Reactor

The sodium-cooled fast reactor container is an integrated pool structure composed of numerous internal components and complex structure. The anisotropy is obvious and the deep penetration problem is serious in the process of neutron transport from core to biological shielding. The calculation of three-dimensional S_N method in large scale is the bottleneck restricting in the design of fast reactor shielding. By combining with high performance computing technology, the parallel computing scheme is used to solve the anisotropic three-dimensional deep penetration shielding calculation in the fast reactor. In this paper, the China Demonstration Fast Reactor (CFR600) reactor block was taken as the research object. Using JSNT-CFR code, the neutron flux rate, photon flux rate, and dose rate in the reactor block were calculated in detail. The calculation results were compared with those of the existing two-dimensional code. The results show that combining the traditional shielding calculation method with high performance computing can meet the requirements of CFR600 reactor block shielding calculation accuracy, and obtain a more comprehensive three-dimensional display effect. It can solve the problem of shielding calculation of complex problems such as complex model and particle penetration depth. It has obvious advantages and provides strong support for the large sodium-cooled fast reactor shielding design.     

Computational tracking of BN-600 operation

Computational tracking of BN-600 operation is described. The high quality of computational tracking is largely due to the nature of a fast reactor, in this case BN-600. Unlike reactors with a thermal neutron spectrum, in a fast reactor, because the prompt and delayed fission neutrons as well as the absorbed neutrons are almost in the same energy range as the fast neutrons, a computational cell can be confidently homogenized and the reactor is strongly coupled to the neutron field. These are the reasons why the behavior of the reactor can be successfully predicted by means of computational programs which are based on the diffusion approximation neglecting the anisotropy of the interaction of the neutrons and the heterogeneity of the medium.     

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型,瞬态热工水力及安全特性是其设计研发和安全评审的重要工作,需要专用的分析工具。本文基于模块化建模思想,建立了钠冷快堆系统关键部件的热工水力模型和辅助模型,采用具有高稳定性和自动变步长能力的Gear算法,开发了钠冷快堆瞬态热工水力及安全分析软件THACS,并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明,THACS程序能较好模拟此实验的瞬态过程,具备钠冷快堆瞬态热工水力及安全分析的能力,可为我国钠冷快堆研发提供支持。     

Experience in Operating the BN-600 Unit at the Belyi Yar Nuclear Power Plant

Data on the experience in operating the BN-600 reactor with sodium coolant in the first and second loops are presented. It is shown that the equipment and the systems in the sodium loops, which have operated for more than 23 years, are highly realiable. The average installed capacity utilization factor for this period reached about 74%. The losses due to rupture of the heat-transfer pipes in the steam-generator modules and the sodium leaks from the loops were about 0.3%. Disruptions of normal operation are detected reliably and contained by the safety systems present in the unit. Unique experience in performing maintenance and repair work on the sodium equipment has been gained on the BN-600 reactor.     

Successive development phases for sodium-cooled fast reactors

The development and operation of sodium-cooled fast reactors and the prospects for developing the next generation of such reactors are reviewed. The main phases, the problems of each phase, and the results obtained by solving these problems are shown. The main results obtained by adopting innovative technical design solutions, making it possible to consider the problem of developing a competitive power-generating unit with a BN-1200 reactor, are examined and described.     

In vessel neutron instrumentation for sodium-cooled fast reactors: Type,lifetime and location

The in vessel instrumentation of sodium-cooled fast reactors must deliver measurements that are reliable and easy to interpret over several reactor cycles in order to fulfill the safety requirements. This paper compares, with respect to this requirement, three types of detectors that are widely used in neutron measurements: fission chambers, boron-lined proportional counters, self-powered neutron detectors. We use neutron spectra that are computed for preliminary design of sodium-cooled fast reactor in different representative locations: in diluting tubes within nuclear fuel assemblies, or in the lateral neutron protections. With an evolution code, we compute the expected signal for each type of detector, to assess whether its level is sufficient, and also its evolution over three operating cycles, to examine whether it is compatible with long term measurements. The conclusion is that fission chambers are the only type able to deliver an interpretable signal for a wide dynamic of reactor power and for three or more operating cycles. The two other types are shown to be inadequate.     

Concept of an advanced power-generating unit with a BN-1200 sodium-cooled fast reactor

The status of work on the development of a 1200 MW sodium-cooled reactor facility for serial construction is presented. The general characteristics of the facility and the power-generating unit as well as the objectives which must be attained as a result of the design are presented. The design of the power-generating group is based on solutions some of which have been checked during the operation of sodium-cooled reactors in Russia and some have been validated by the appropriate research and development work performed for BN-800. At the same time, new solutions are used which are aimed at improving the technical-economic indicators and increasing the level of safety. Additional R&D work will be needed to validate them.     

Improvement of the equipment in fast-neutron reactor facilities

Information is presented on the development of the main equipment of the BN-1200 advanced reactor facility: first- and second-loop main circulation pumps, intermediate heat exchanger, actuation mechanism of the cold filter-trap control and protection system, autonomous and air heat exchangers, steam generator. The approach to the development of the equipment is based on maximum use of the experience gained in operating BN-350 and -600 as well as experience in developing the BN-800 design, which gives a basis for ensuring reliable operation of the BN-1200 equipment. New solutions for improving the technical-economic indicators and increasing the safety of a power-generating unit, whose validation required R&D work, are examined at the same time.     

钠冷快堆换料系统可靠性研究

由于中子通量以及冷却剂运行温度高,钠冷快中子反应堆（简称钠冷快堆）的换料周期较一般轻水反应堆短。同时,换料过程中隔绝空气的要求以及换料设备本身的复杂性,钠冷快堆只能逐根进行换料,使得总的换料时间较轻水反应堆长。本文采用失效模式与影响分析、故障树分析等方法对典型钠冷快堆换料系统各部分的可靠性进行评价,获得了换料系统每次换料期间的失效概率。基于换料系统各部分失效的影响、失效概率以及恢复时间,分析了换料系统不同失效模式对反应堆运行效率的影响。     

Development and validation of a fast reactor core burnup code – FARCOB

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X–Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.     

BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.     

应用于钠冷快堆的超临界二氧化碳动力转换系统研究

超临界二氧化碳（SCO₂）布雷顿循环由于高效、紧凑和可避免钠水反应等特性而成为钠冷快堆的理想动力转换系统。本文以1 200 MWe大型池式钠冷快堆为系统热源,钠回路温度及热负荷为循环系统运行边界,对比研究了不同SCO₂布雷顿循环系统性能和关键设备性能的变化规律。研究发现,级间冷却再压缩循环与钠冷快堆热源特性匹配性最佳,且循环效率最高（40.7%）。进而研究了不同运行参数对级间冷却再压缩循环效率的影响规律,给出了循环系统效率对各关键影响因素的敏感度,发现循环系统效率对冷端参数的敏感度最强,其次为分流比和透平入口参数,对主压缩机级间压比的敏感度最弱。     

Supercritical Carbon Dioxide Power Conversion System Applied to Sodium-cold Fast Reactor

Because of the high efficiency, compactness and avoiding sodium water reaction, the supercritical carbon dioxide (SCO₂) Brayton cycle is an ideal power conversion system for sodium-cooled fast reactors. In this paper, the 1 200 MWe Sodium-cooled Fast Reactor was used as the heat source of the system, and the temperature and heat load of the sodium loop were used as the operating boundary of the circulation system. The system performance and key equipment performance of different supercritical carbon dioxide Brayton cycles were compared. The coupling between the inter-stage cooling and recompression cycle and the characteristics of the heat source of the sodium-cooled reactor is the best, and the cycle efficiency is the highest (40.7%). Furthermore, the influence of different operating parameters on the efficiency of the inter-stage cooling and recompression cycle was studied, and the sensitivity of the efficiency of the circulation system to each of the key influencing factors was given. It is found that the efficiency of the circulation system is the most sensitive to the cold-end parameters, followed by the split ratio and turbine inlet parameters, and the weakest to the main compressor inter-stage pressure ratio.     

Demonstration Experiment of 3 BN-600 MOX Vibropac FAs Irradiation for the Excess Weapons Plutonium Disposal

The paper presents results of a demonstration experiment on conversion of 50 kg of weapon-grade plutonium in the form of metal ingots into granulated MOX-fuel to be used for manufacturing fuel pins and 3 fuel assemblies (FAs) for the fast power-generating reactor BN-600, irradiation parameters of these FAs and the data from post-irradiation examinations. It can be concluded from the PIE results that the 3FAs were successfully irradiated in BN-600 without any fuel pin failures. Therefore, disposition of weapongrade plutonium with a weight of about 20 kg was successfully done. This represents the first disposition of Russian surplus weapon-grade plutonium as an international cooperation (this experiment was performed in collaboration between RIAR and JNC). The possibility of using MOX vipac fuel as a method for weapon plutonium disposition is clearly shown.